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DOE OSTI2017

Preliminary Evaluation of Removing Used Nuclear Fuel from Shutdown Sites

A preliminary evaluation of removing spent nuclear fuel (SNF) from 14 shutdown nuclear power plant sites was performed. At these shutdown sites the nuclear power reactors have been permanently shut down and the sites have been decommissioned or are undergoing decommissioning. The shutdown sites were Maine Yankee, Yankee Rowe, Connecticut Yankee, Humboldt Bay, Big Rock Point, Rancho Seco, Trojan, La Crosse, Zion, Crystal River, Kewaunee, San Onofre, Vermont Yankee, and Fort Calhoun. The evaluation was divided into four components: (1) characterization of the SNF and greater-than-Class C low-level radioactive waste (GTCC waste) inventory, (2) a description of the on-site infrastructure at the shutdown sites, (3) an evaluation of the near-site transportation infrastructure and transportation experience at the shutdown sites, and (4) an evaluation of the actions necessary to prepare for and remove SNF and GTCC waste. The primary sources for the inventory of SNF and GTCC waste were the U.S. Department of Energy (DOE) spent nuclear fuel inventory database, industry publications such as StoreFUEL, and government sources such as the U.S. Nuclear Regulatory Commission. The primary sources for information on the conditions of on-site infrastructure and near-site transportation infrastructure and experience included information collected during site visits, information provided by managers at the shutdown sites, Facility Interface Data Sheets compiled for DOE in 2005, Services Planning Documents prepared for DOE in 1993 and 1994, industry publications such as Radwaste Solutions, and Google Earth. State representatives, State Regional Group representatives, Tribal representatives, and a Federal Railroad Administration representatives have participated in eleven of the shutdown site visits. Every shutdown site was found to have at least one off-site transportation mode option for removing its SNF and GTCC waste; some have multiple options. Experience removing large components during reactor decommissioning provided an important source of information used to identify the transportation mode options for the sites. Especially important in conducting the evaluation were site visits, through which information was obtained that would not have been available otherwise. Extensive photographs taken during the site visits proved to be particularly useful in documenting the current conditions at or near the sites. Additional conclusions from this evaluation include: The 14 shutdown sites use designs from 4 different suppliers involving 11 different (horizontal and vertical) dry storage systems that would require the use of 9 different transportation cask designs to remove the SNF and GTCC waste from the shutdown sites. Although some changes to transportation certificates of compliance will be required, the SNF at the initial 9 shutdown sites (Maine Yankee, Yankee Rowe, Connecticut Yankee, Humboldt Bay, Big Rock Point, Rancho Seco, Trojan, La Crosse, and Zion) is in dual purpose dry storage canisters that can be transported, including a small amount of high-burnup fuel. Most sites indicated that 2-3 years of advance time would be required for its preparations before shipments could begin. Some sites could be ready in less time. As additional sites such as Palisades, Pilgrim, Three Mile Island, Oyster Creek, Indian Point, and Diablo Canyon shut down, these sites will be included in updates to the evaluation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

Measurement of radiolytic hydrogen generation and impact of drying treatments on reactor exposed and surrogate aluminum materials

Technical challenges associated with dry storage of Aluminum-clad Spent Nuclear Fuel (ASNF) include a need to better understand the potential extent of gaseous molecular hydrogen (H 2 ) production through radiolytic degradation of the aluminum (oxy)hydroxide films present on ASNF. The characterization of radiolytic gas generation from ASNF (oxy)hydroxide layers has been identified as a key knowledge gap which poses a technical challenge to the long-term storage of ASNF. Task 2 of the action plan for the extended (>50 years) dry storage of ASNF addresses this gap. Previous radiolysis studies performed under Task 2 established baseline estimates of H 2 generation rates from the attendant hydrated oxides. The results confirmed that net radiolytic H 2 production has a dependency on absorbed gamma dose, as well as relative humidity and cover gas composition (air, nitrogen, and argon). Further experiments revealed that the physisorbed water on the samples may significantly impact the radiolytic H 2 yield. This phenomenon complicates the determination of H 2 generation rates for hydrated oxides, which may lead to inaccurate modeling predictions of the long-term H 2 yields in sealed storage systems containing ASNF, particularly when compared to spent fuel casks which have undergone some drying process intended to remove physically- and chemically-bound water. This report describes the testing methods utilized and the hydrogen generation results obtained in an investigation of the effects of gamma irradiation on aluminum materials for a variety of (oxy)hydroxide surface compositions and drying conditions. The testing methods included small-area aluminum material testing in ampules, and large-area aluminum material testing in steel vessels. Test material preparation, irradiation, and radiolytic H 2 measurement methods are summarized. The measured H 2 concentrations, which reflect the variable initial hydrated inventory and drying treatments, are compared to one another as well as to previously published data, to identify the primary factors affecting radiolytic H 2 generation rates and potential equilibrium H 2 concentrations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Blind Modeling Validation Exercises Using the Horizontal Dry Cask Simulator

The U.S. Department of Energy (DOE) established a need to understand the thermal-hydraulic properties of dry storage systems for commercial spent nuclear fuel (SNF) in response to a shift towards the storage of high-burnup (HBU) fuel (> 45 gigawatt days per metric ton of uranium, or GWd/MTU). This shift raises concerns regarding cladding integrity, which faces increased risk at the higher temperatures within spent fuel assemblies present within HBU fuel compared to low-burnup fuel (≤ 45 GWd/MTU). A dry cask simulator (DCS) was built at Sandia National Laboratories (SNL) in Albuquerque, New Mexico to produce validation-quality data that can be used to test the accuracy of the modeling used to predict cladding temperatures. These temperatures are critical to evaluating cladding integrity throughout the storage cycle of commercial spent nuclear fuel. A model validation exercise was previously carried out for the DCS in a vertical configuration. Lessons learned during the previous validation exercise have been applied to a new, blind study using a horizontal dry cask simulator (HDCS). Three modeling institutions – the Nuclear Regulatory Commission (NRC), Pacific Northwest National Laboratory (PNNL), and Empresa Nacional del Uranio, S.A., S.M.E. (ENUSA) – were granted access to the input parameters from the DCS Handbook, SAND2017-13058R, and results from a limited data set from the horizontal BWR dry cask simulator tests reported in the HDCS update report, SAND2019-11688R. With this information, each institution was tasked to calculate peak cladding temperatures and air mass flow rates for ten HDCS test cases. Axial as well as vertical and horizontal transverse temperature profiles were also calculated. These calculations were done using modeling codes (ANSYS/Fluent, STAR-CCM+, or COBRA-SFS), each with their own unique combination of modeling assumptions and boundary conditions. For this validation study, the ten test cases of the horizontal dry cask simulator were defined by three independent variables – fuel assembly decay heat (0.5 kW, 1 kW, 2.5 W, and 5 kW), internal backfill pressure (100 kPa and 800 kPa), and backfill gas (helium and air). The plots provided in Chapter 3 of this report show the axial, vertical, and horizontal temperature profiles obtained from the dry cask simulator experiments in the horizontal configuration and the corresponding models used to describe the thermal-hydraulic behavior of this system. The tables provided in Chapter 3 illustrate the closeness of fit of the model data to the experiment data through root mean square (RMS) calculations of the error in peak cladding temperatures (PCTs), PCT axial locations, axial temperature profiles, vertical and horizontal temperature profiles at two different axial locations, and air mass flow rates for the ten test cases, normalized by the experimental results. The model results are assigned arbitrary model numbers to retain anonymity. Due to the relatively flat axial temperature profiles, small temperature gradients resulted in large deviations of all models’ PCT axial location from the experimental PCT axial location. When the PCT axial location error is excluded in the calculation of the combined RMS of the normalized errors that considers PCT, the temperature profiles, and the air mass flow rates, the model data fits the experimental data to within 5%. When the vault information is excluded, the model data fits the experimental data to within 2.5%. An error analysis was developed further for one model, using the model and experimental uncertainties in each validation parameter to calculate validation uncertainties. The uncertainties for each parameter were used to define quantifiable validation criteria. For this analysis, the model was considered validated for a given comparison metric if the normalized error in that metric divided by the validation uncertainty was less than or equal to 1. When considering the combined RMS of the normalized errors of all metrics divided by their validation uncertainties, the model was found to have satisfied the criterion for model validation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2018

Post-irradiation Examinations of High Burnup PWR Rods - 18456

The High Burnup Spent Fuel Data Project (herein referred to as the 'project'), sponsored by the US Department of Energy (DOE) Office of Nuclear Energy (NE), is focused on experimentally evaluating the effects of long-term storage and transportation on high burnup (HBU) (defined as >45 gigawatt days per metric ton uranium) pressurized water reactor (PWR) spent nuclear fuel (SNF) [1]. To support this project, in the fall of 2017, an instrumented bolted-lid cask (project cask) at the North Anna reactor site was loaded with intact, HBU PWR fuel assemblies (project assemblies) having four different kinds of cladding and will be reopened in approximately 10-years for inspection. To support pre- and post-storage SNF rod characterization, 25 'sister rods' that are similar to the rods being stored in the project cask were extracted from seven different fuel assemblies at the North Anna nuclear power plant for post-irradiation examination (PIE). The sister rod testing program has multiple objectives, including: 1) Baseline testing to provide essential information on the physical condition and properties of the cladding and the fuel in the cladding prior to the loading, drying, and long-term dry storage process (analogous to an experiment control group). Similar tests are expected to be performed at the end of the 10-year storage period on rods extracted from the cask to identify any changes in the properties of the fuel rods that may have occurred during the dry storage period; and 2) Tests to provide material property and physical data using accelerated separate effects tests (SETs) and small-scale tests (SSTs) to explore a wide range of conditions that may occur for other dry storage systems and/or while the fuel is being transported after aging. Testing will be focused on understanding overall SNF rod strength and durability, and be comprised of both composite fuel and defueled clad testing with a focus on measuring cladding material and mechanical properties. Nondestructive testing is currently underway and plans are being updated to support the initial set of destructive tests that are expected to begin by early 2018. While the data generated can be used for multiple purposes, a primary goal of the sister rod PIE data and associated measured mechanical properties are expected to support existing fuel storage licensing and relicensing activities by addressing identified knowledge gaps, and enhance the technical bases for post-storage transportation, handling, and consolidation activities [1]. This paper presents results obtained from the sister rod testing to date, and describes some of the planned measurements, equipment, and procedures that will be implemented for the initial set of destructive examinations. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2015

Criticality Safety Analysis Of As-loaded Spent Nuclear Fuel Casks

The final safety analysis report (FSAR) or the safety analysis report (SAR) for a particular spent nuclear fuel (SNF) cask system documents models and calculations used to demonstrate that a system meets the regulatory requirements under all normal, off-normal, and accident conditions of spent fuel storage, and normal and accident conditions of transportation. FSAR/SAR calculations and approved content specifications are intended to be bounding in nature to certify cask systems for a variety of fuel characteristics with simplified SNF loading requirements. Therefore, in general, loaded cask systems possess excess and uncredited criticality margins (i.e., the difference between the licensing basis and the as-loaded calculations). This uncredited margin could be quantified by employing more detailed cask-specific evaluations that credit the actual as-loaded cask inventory, and taking into account full (actinide and fission product) burnup credit. This uncredited criticality margin could be potentially used to offset (1) uncertainties in the safety basis that needs to account for the effects of system aging during extended dry storage prior to transportation, and (2) increases in SNF system reactivity over a repository performance period (e.g., 10,000 years or more) as the system undergoes degradation and internal geometry changes. This paper summarizes an assessment of cask-specific, as-loaded criticality margins for SNF stored at eight reactor sites (215 loaded casks were analyzed) under fully flooded conditions to assess the margins available during transportation after extended storage. It is observed that the calculated k eff margin varies from 0.05 to almost 0.3 Δk eff for the eight selected reactor sites, demonstrating that significant uncredited safety margins are present. In addition, this paper evaluates the sufficiency of this excess margin in applications involving direct disposal of currently loaded SNF casks.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

An Evaluation of Standardized Canisters in the Waste Management System - 16317

Development of a standardized canister system represents an opportunity to develop an integrated approach to addressing storage, transportation, and disposal issues in the waste management system. However, regardless of timing and method, deployment of such a system would have the potential to cause significant system-wide impacts. This evaluation expands upon previous standardized-canister system evaluations and compares continued loading of dual-purpose canisters (i.e., the status quo) with loading of standardized canister systems in the near term, that is, before repository requirements are known and/or before operating reactors shut down. This evaluation quantitatively compares order-of-magnitude costs and logistics for different standardization scenarios with those of status quo scenarios, provides insight into quantifiable impacts of loading standardized canister systems in the near term, tests system-level analysis tools and associated input, and identifies scenarios for further analysis. This is a technical paper that does not take into account the contractual limitations under the Standard Contract (10 CFR Part 961). As work on assessing standardized canister system scenarios has progressed, new data for at-reactor and interim storage facility (ISF) operations have been collected and/or generated to provide more realism at the system level, and past scenarios were re-evaluated with this new information. In addition, new scenarios were developed and analyzed that focused on (1) bare fuel transported from reactors to an ISF in re-usable bare fuel casks and deferring deployment of a standardized canister system to the ISF, (2) alternative acceptance strategies based on conclusions from previous systems architecture studies, and (3) the impacts of having to accommodate spent fuel in differently sized canisters for different potential repository geologies. Results of this evaluation indicate that shifting the loading of standardized canisters from the reactor to the ISF does not result in a significant change in total system cost. Results also show that some alternative acceptance strategies could reduce the number of years that spent nuclear fuel (SNF) stays on-site at reactors for cases involving standardized canisters (consistent with previous results that looked at dual-purpose canisters). Finally, incorporating the new at-reactor loading data of small standardized canisters shows a significant system-wide cost reduction when compared to previous estimates of at-reactor loading of small canisters. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2015

Addressing Fission Product Validation in MCNP Burnup Credit Criticality Calculations

The US Nuclear Regulatory Commission (NRC) Division of Spent Fuel Storage and Transportation issued Interim Staff Guidance (ISG) 8, Revision 3 in September 2012. This ISG provides guidance for NRC staff members’ review of burnup credit (BUC) analyses supporting transport and dry storage of pressurized water reactor spent nuclear fuel (SNF) in casks. The ISG includes guidance for addressing validation of criticality (k eff ) calculations crediting the presence of a limited set of fission products and minor actinides (FP&MAs). Based on previous work documented in NRC Regulatory Guide (NUREG) Contractor Report (CR)-7109, the ISG recommends that NRC staff members accept the use of either 1.5 or 3% of the FP&MA worth—in addition to bias and bias uncertainty resulting from validation of k eff calculations for the major actinides in SNF—to conservatively account for the bias and bias uncertainty associated with the specified unvalidated FP&MAs. The ISG recommends (1) use of 1.5% of the FP&MA worth if a modern version of SCALE and its nuclear data are used and (2) 3% of the FP&MA worth for well qualified, industry standard code systems other than SCALE with the Evaluated Nuclear Data Files, Part B (ENDF/B),-V, ENDF/B-VI, or ENDF/B-VII cross sections libraries. The work presented in this paper provides a basis for extending the use of the 1.5% of the FP&MA worth bias to BUC criticality calculations performed using the Monte Carlo N-Particle (MCNP) code. The extended use of the 1.5% FP&MA worth bias is shown to be acceptable by comparison of FP&MA worths calculated using SCALE and MCNP with ENDF/B-V, -VI, and -VII–based nuclear data. The comparison supports use of the 1.5% FP&MA worth bias when the MCNP code is used for criticality calculations, provided that the cask design is similar to the hypothetical generic BUC-32 cask model and that the credited FP&MA worth is no more than 0.1 Δk eff (ISG-8, Rev. 3, Recommendation 4).

42 ENGINEERING↗

DOE OSTI2022

The Hanford Lead Canister Collaboration

The Hanford Site, Management of the Cesium and Strontium Capsules (MCSC) Project is preparing to place cesium and strontium material into dry storage using stainless steel canisters that are housed inside vertical concrete casks. Part of the site’s aging management program is to use a spare canister system as a leading indicator canister to provide advance warning for signs of chloride induced stress corrosion cracking (CISCC) or any other kind of canister degradation that might take place over the system’s 300-year design life. The Hanford Lead Canister (HLC) will use electric heaters to simulate the decay heat of cesium and strontium to ensure the HLC simulates the environment and conditions the capsule-loaded systems will experience. In addition to being part of the site’s aging management program, the HLC is being made available for research and development that advances the industry regarding canister mitigation and repair (M&R). The HLC system closely resembles current spent nuclear fuel (SNF) dry storage systems, making the HLC useful for demonstrating technologies or performing research that is relevant to SNF canisters . Several organizations are collaborating to prepare the HLC, along with technical strategy, to perform the needed research and development activities, to include the Pacific Northwest National Laboratory (PNNL), Central Plateau Cleanup Company (CPCCo), and the Electric Power Research Institute (EPRI). In one example of this collaborative research, EPRI is planning to use the HLC to demonstrate in-situ M&R technologies under controlled conditions before the HLC is deployed at Hanford. There is a period of several years when the HLC will be available for canister M&R activity before it is deployed. This paper describes the development of the HLC, its role in the Hanford site’s MCSC Project aging management program, and some of the research activities that are planned for the HLC.

Klymyshyn, Nicholas A.↗

DOE OSTI2018

Potential Research and Development Opportunities for Light Water Reactor Spent Nuclear Fuel at INL

This report documents the commercial spent nuclear fuel currently in storage at Idaho National Laboratory with a specific focus on the potential research and development (R&D) opportunities for this material. Commercial spent nuclear fuel is stored in dry casks on the INL site at the Idaho Nuclear Technology and Engineering Center (INTEC, formerly the Idaho Chemical Processing Plant or ICPP). This spent fuel was part of the DOE-industry cooperative dry cask demonstration project that began in 1983 and from historical fuel testing and treatment activities that had been performed at the Test Area North (TAN)-607 facility. Most of the fuel is “low burnup,” but given that it was removed from the respective reactors between 1970 and 1985, it was representative for its time. The majority of the fuel is of the 15×15 Westinghouse PWR variety, though additional fuel types (e.g., TN-BRP) and configurations are also stored. Potential R&D opportunities identified are (1) the REA-2023 cask, (2) the TN-BRP cask, and (3) the Castor V/21 cask. Fuel from the TN-24P may be an alternative to the intact assemblies in the Castor V/21. Specifically, the REA-2023 cask contains 9 rods that have been exposed to air since 2005. These rods have a detailed pre-loading characterization and thus any changes in corrosion or geometry can most likely be attributed to the storage environment. In addition, similar rods to the nine in REA-2023 reside in the Castor V/21 cask whose environment has been maintained with inert nitrogen. The Castor V/21 cask also contains intact PWR assemblies that could be used to perform limited confirmation of the pyroprocessing approach. An alternative to the use of the Castor assemblies for pyroprocessing is the consolidated fuel that is stored in the TN-24P, which would not require removal from the assembly grid since it has already been packaged in accessible two-piece shell strongback containers. The TN-BRP cask contains a number of Gd-doped, low burnup assemblies that could be of use to validate code packages used to justify burnup-up credit for boiling water reactors (BWRs). The INL dry casks are stored on the CPP-2707 pad at INTEC, adjacent to the CPP-603 Irradiated Fuel Storage Facility, which incorporates a dry hot cell that has the potential to receive a full scale commercial cask of the types identified. Several evaluations have been made regarding this task, including full scale mockup testing of the physical clearances required to remove and handle fuel assemblies and components. A proposed schedule of 14 months and cost of $5.3 million was the outcome of analysis performed by Fluor Idaho and is included as Appendix A & B. In addition to the spent fuel at INL, there is also a full-scale mockup assembly that was fabricated for the Protypical Consolidation Demonstration Project to validate fuel rod consolidation approaches. The assembly was fabricated by Westinghouse as a 15 x 15 PWR, using Zircaloy-4 tubing and steel end fittings in accordance with product specifications. The only departure from specification is the use of copper metal in lieu of uranium oxide pellets to achieve a weight that is within 10% of the standard.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

TN{sup R} MW- Multi Waste Package 'All In One Solution' For Conditioning - Transport - Storage - Up To Final Disposal - 17196

For the management of radioactive waste, either during nuclear facility operation or decommissioning, one must consider a diverse array of nuclear waste forms of various types, volumes and activities. Examples are: activated metallic waste; legacy waste; sludge; resins; and orphan waste. Customers are frequently faced with the obligation of undertaking multiple, costly waste management operations. These operations activities include handling, reconditioning or material transfers between packages, as in the case when moving material from on-site storage to transportation, and then from transportation to final disposal. To address this issue, a new - highly flexible up to reversible - cask system, the TN{sup R} MW, is being developed as an 'All in One Solution'. TN{sup R} MW has a total weight of 10 metric tons and is compliant with the 2012 International Atomic Energy Agency (IAEA) regulations. It was developed to provide flexibility and designed with the capability to be adaptable to the various needs of the nuclear industry. From IP-2 and Type A containers to Type B(U) and B(U)F casks, this cask offers different options to sufficiently address the requirements of each step of the waste stream from on-site and international transport to long-term interim storage and final disposal. Licensing and manufacturing of a number of packagings in this TN{sup R} MW family are underway. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2022

MELCOR Input Model for Spent Fuel Transportation and Storage Canister

This report documents the progress and current results of the MELCOR spent fuel cask input model. The MELCOR model is being developed to investigate aerosol transport and deposition given the aerosol physical phenomena models within MELCOR. To perform the analyses, a general portrayal of the MAGNASTOR® cask system has been employed; however, this system was selected based on available information to provide a reasonable representation of a spent fuel cask. The analytical results are not intended to characterize the performance of the MAGNASTOR® cask. Instead, the provided results are intended to enhance our general understanding of the aerosol behavior within casks and the validity of current models. The current model efforts are being performed to investigate hypothetical UO 2 release from failed fuel pins within a spent fuel cask. The existing MELCOR model of the MAGNASTOR® cask system has been adapted to permit future comparative analyses with the GOTHIC representation of the MAGNASTOR® cask. To support this comparison, the PNNL model characteristics that are unrelated to the aerosol modeling were applied to the MELCOR model. These characteristics included improved comparability of the axial fidelity, total spent fuel power, fuel pin axial power profile, and heat losses from cannister. The thermal-hydraulic solutions are improved within the capability of the MELCOR code and will permit better overall agreement with the GOTHIC results. Detailed results are presented on the thermal-hydraulic analysis of the MELCOR cask as well as characterization of UO 2 aerosol dispersion and deposition within the cask.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2015

Criticality safety analysis of as-loaded spent nuclear Fuel Casks - 14589

The final safety analysis report (FSAR) or the safety analysis report (SAR) for a particular spent nuclear fuel (SNF) cask (both bare fuel and canister-based) system documents models and calculations used to demonstrate that a system meets the regulatory requirements under all normal, off-normal, and accident conditions of spent fuel storage, as well as the normal and accident conditions of transportation. FSAR/SAR calculations and approved content specifications are intended to be bounding in nature to certify cask systems for a variety of fuel characteristics with simplified SNF loading requirements. Therefore, in general, loaded cask systems possess excess and uncredited criticality margins (i.e., the difference between the licensing basis and the as-loaded calculations). This uncredited margin could be quantified by (1) employing more detailed cask-specific evaluations that credit the actual as-loaded cask inventory and (2) taking into account full (actinide and fission product) burnup credit. This uncredited criticality margin could be used to offset uncertainties in the safety basis to account for the effects of system aging during extended dry storage prior to transportation. Additionally, the uncredited criticality margin could also potentially be credited to offset increases in SNF system reactivity that may occur over a repository performance period (e.g., 10,000 years or more) as the system undergoes degradation and internal geometry changes. This paper summarizes an assessment of cask-specific as-loaded criticality margins for 215 loaded casks of SNF stored at eight reactor sites under fully flooded conditions to assess the margins available during transportation after extended storage. It is observed that the calculated k{sub eff} margin varies from 0.05 to almost 0.3 Δk{sub eff} for the eight selected reactor sites, demonstrating that significant uncredited safety margins are present. In addition, this paper evaluates the sufficiency of this excess margin in applications involving direct disposal of currently loaded SNF canisters (for canister-based systems) in a repository. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2014

Preliminary Evaluation of Removing Used Nuclear Fuel from Shutdown Sites

This report presents a preliminary evaluation of removing used nuclear fuel (UNF) from 12 shutdown nuclear power plant sites. At these shutdown sites the nuclear power reactors have been permanently shut down and the sites have been decommissioned or are undergoing decommissioning. The shutdown sites are Maine Yankee, Yankee Rowe, Connecticut Yankee, Humboldt Bay, Big Rock Point, Rancho Seco, Trojan, La Crosse, Zion, Crystal River, Kewaunee, and San Onofre. The evaluation was divided into four components: characterization of the UNF and greater-than-Class C low-level radioactive waste (GTCC waste) inventory; a description of the on-site infrastructure and conditions relevant to transportation of UNF and GTCC waste; an evaluation of the near-site transportation infrastructure and experience relevant to shipping transportation casks containing UNF and GTCC waste, including identification of gaps in information; and, an evaluation of the actions necessary to prepare for and remove UNF and GTCC waste. The primary sources for the inventory of UNF and GTCC waste are the U.S. Department of Energy (DOE) RW-859 used nuclear fuel inventory database, industry sources such as StoreFUEL and SpentFUEL, and government sources such as the U.S. Nuclear Regulatory Commission. The primary sources for information on the conditions of site and near-site transportation infrastructure and experience included observations and information collected during visits to the Maine Yankee, Yankee Rowe, Connecticut Yankee, Humboldt Bay, Big Rock Point, Rancho Seco, Trojan, La Crosse, and Zion sites; information provided by managers at the shutdown sites; Facility Interface Data Sheets compiled for DOE in 2005; Services Planning Documents prepared for DOE in 1993 and 1994; industry publications such as Radwaste Solutions; and Google Earth. State and Regional Group representatives, a Tribal representative, and a Federal Railroad Administration representative participated in six of the shutdown site visits. Every site was found to have at least one off-site transportation mode option for removing its UNF and GTCC waste; some have multiple options. Experience removing large components during reactor decommissioning provided an important source of information used to identify the transportation mode options for the sites. Especially important in conducting the evaluation were site visits, through which information was obtained that would not have been available otherwise. Extensive photographs taken during the site visits proved to be particularly useful in documenting the current conditions at or near the sites. Additional conclusions from this evaluation include: The 12 shutdown sites use designs from 4 different suppliers involving 9 different (horizontal and vertical) dry storage systems that would require the use of 8 different transportation cask designs to remove the UNF and GTCC waste from the shutdown sites; Although there are common aspects, each site has some unique features and/or conditions; Although some regulatory actions will be required, all UNF at the initial 9 shutdown sites (Maine Yankee, Yankee Rowe, Connecticut Yankee, Humboldt Bay, Big Rock Point, Rancho Seco, Trojan, La Crosse, and Zion) is in licensed systems that can be transported, including a small amount of high-burnup fuel; Each site indicated that 2-3 years of advance time would be required for its preparations before shipments could begin; Most sites have more than one transportation option, e.g., rail, barge, or heavy haul truck, as well as constraints and preferences. It is expected that additional site visits will be conducted to add to the information presented in the evaluation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2015

Beyond 5% enrichment limit for PWR 17x17 and VVER-440 fuel - Criticality problems - 14207

The fuel suppliers improve permanently fuel for commercial energetic nuclear reactors. One way is to increase initial enrichment of U{sup 235}. Higher initial enrichment of U{sup 235} allows to increase burnup of assembly, to prolong stay in assembly in reactor and therefore decrease number of fresh assemblies loaded into reactor and decrease number of spent fuel assemblies for storage and disposal. In this article are some criticality analyses for fresh fuel PWR 17x17 and VVER-440 for enrichment up to 6% of U{sup 235} without burnable absorber: - Criticality of fresh assembly (PWR 17x17 and VVER-440) in cold state in reactor - Critical number of fresh assemblies (PWR 17x17 and VVER-440) in water - Possible storage and transport of fresh fuel VVER-440 in existing facilities and transport cask The SCALE 6.1.2 system was used for all calculation. The KENO VI module was used with library v7-238 (continuous energy). (author)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2018

Post-irradiation Examinations of High Burnup PWR Rods

The High Burnup Spent Fuel Data Project (herein referred to as the “project”), sponsored by the US Department of Energy (DOE) Office of Nuclear Energy (NE), is focused on experimentally evaluating the effects of long-term storage and transportation on high burnup (HBU) (defined as >45 gigawatt days per metric ton uranium) pressurized water reactor (PWR) spent nuclear fuel (SNF) [1]. To support this project, in the fall of 2017, an instrumented bolted-lid cask (project cask) at the North Anna reactor site was loaded with intact, HBU PWR fuel assemblies (project assemblies) having four different kinds of cladding and will be reopened in approximately 10-years for inspection. To support pre- and post-storage SNF rod characterization, 25 “sister rods” that are similar to the rods being stored in the project cask were extracted from seven different fuel assemblies at the North Anna nuclear power plant for post-irradiation examination (PIE).The sister rod testing program has multiple objectives, including: 1) Baseline testing to provide essential information on the physical condition and properties of the cladding and the fuel in the cladding prior to the loading, drying, and long-term dry storage process (analogous to an experiment control group). Similar tests are expected to be performed at the end of the 10-year storage period on rods extracted from the cask to identify any changes in the properties of the fuel rods that may have occurred during the dry storage period; and 2) Tests to provide material property and physical data using accelerated separate effects tests (SETs) and small-scale tests (SSTs) to explore a wide range of conditions that may occur for other dry storage systems and/or while the fuel is being transported after aging.Testing will be focused on understanding overall SNF rod strength and durability, and be comprised of both composite fuel and defueled clad testing with a focus on measuring cladding material and mechanical properties. Nondestructive testing is currently underway and plans are being updated to support the initial set of destructive tests that are expected to begin by early 2018. While the data generated can be used for multiple purposes, a primary goal of the sister rod PIE data and associated measured mechanical properties are expected to support existing fuel storage licensing and relicensing activities by addressing identified knowledge gaps, and enhance the technical bases for post-storage transportation, handling, and consolidation activities [1]. This paper presents results obtained from the sister rod testing to date, and describes some of the planned measurements, equipment, and procedures that will be implemented for the initial set of destructive examinations.

Montgomery, Rose↗

DOE OSTI2023

Sister Rod Destructive Examinations (FY23) Appendix F3: Uncertainty and Conservative Bias in the Cyclic Integrated Reversible-Bending Fatigue Test

The development and application of the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT) has been documented over the past 10 years. The CIRFT system was developed in collaboration with the U.S. Nuclear Regulatory Commission (NRC) to address the concern that high-burnup fuel rod fatigue performance was degraded compared to cladding-only behavior. Fatigue data measured by CIRFT on high-burnup fuel rods were then used by the NRC to develop a fatigue response curve for pressurized water reactor fuel rods with Zircaloy-4 cladding in NUREG-2224. Subsequently, CIRFT has been used in the sister rod project to test whether reoriented hydrides associated with vacuum drying influenced the fatigue lifetime, to extend the fatigue testing database to include fuel rods with ZIRLO cladding, and to add to the experience on both Zircaloy-4 and M5 cladding. These data will be used as a reference for subsequent characterization of fuel rods removed from the high-burnup demonstration cask to better understand the effects of long-term dry storage, which can be thought of as a multiyear elevated temperature anneal. The objective of fatigue testing is to provide data that can be used to develop a best-estimate fatigue curve and a conservative fatigue design curve that accounts for uncertainty and can be compared with actual transportation conditions. The process for defining a design fatigue curve for commercial used fuel rods was developed by Oak Ridge National Laboratory (ORNL) in FY22. One issue noted in the fatigue evaluation was that the calculated combined uncertainty in the strain amplitude values determined from the measured data was large. It was proposed then to directly measure the uncertainty in the strain amplitude to either validate the calculated combined uncertainty or update the uncertainty based on the measurement data. Thus, the purpose of this appendix is to document the measured uncertainty in the strain amplitude data. In the process, it was determined that there has been a conservative bias in the calculated strain amplitude that is small enough to be within the measurement uncertainty. This observation is discussed herein, and a case is made to continue to include the conservative bias in the data.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2015

Development of the computer code for multi-calculations of source term of spent fuel - 15351

Many PWR spent fuels have been transferred between the neighboring nuclear power plants at Kori, Hanbit, Hanwul site, and from nuclear power plant to the KAERI(Korea Atomic Energy Research Institute) for the research purpose in Korea. In case of the PHWR type spent fuels, these have been transferred from the spent fuel pool to an on-site dry storage facility. Total radioactivity of spent fuels has to be calculated before transport in advance by the regulation requirement. In case of PWR, KN-12 transport cask can transfer 12 spent fuel assemblies at one time and in case of PHWR, Hi-STAR 63 transport cask can transfer 120 spent fuel assemblies at one time. Therefore, total radioactivity of all transported spent fuel assemblies that are transported generally are calculated by using the ORIGEN-ARP to meet this regulatory requirement. Basically, ORIGEN-ARP calculates the one fuel assembly at one time. SCALE code including this code provides the batch mode to perform the multi-calculations of the source term. In this study, in order to use the batch mode of ORIGEN-ARP, ARP-auto program was developed to generate input files and batch file of ORIGEN-ARP automatically and to make the summary file extracting the result from output files of ORIGEN-ARP. This code enhances the efficiency at the field. The working time to complete the calculation of source terms of 120 spent fuel assemblies take only 20 minutes. In order to confirm the credibility of this code, total radioactivity of 12 spent fuel assemblies of 17x17 Westinghouse type were calculated and compared to the previous result. The initial enrichments of these spent fuels are 3.46 ∼ 3.50 w/o and burnups are 29 ∼ 43 GWD/MTU. As the results, the relative errors are about 0.9 ∼ 3.0%. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2015

Comparison of German and American fabrication standards for transportation and storage containments of high level radioactive material type B packages - 15055

As the competent authority for the evaluation of German radioactive material (RAM) transport and storage packages, BAM (the German Federal Institute for Materials Research and Testing) has evaluated various international cask manufacturers regarding compliance with the German Type B(U) licensing requirements. The US NRC has recommended paragraphs in the ASME Boiler and Pressure Vessel (B and PV) Code, Section III, Division 3, as the measure to quantify the safety requirements of the US regulations regarding fabrication of RAM transportation and storage containments. The objective of this study is to compare the fabrication criteria of the German Dangerous Goods Procedural Rules (BAM-GGR) and the German Institute for Standardization (DIN) with the US quantitative measures for fabrication of RAM package containment vessels. This comparison will cover major similarities and differences between the requirements of the German standards and the ASME Code requirements for Type B package containment fabrication QC (quality control). While the material most often used for German RAM containment applications is ductile cast iron (DCI), in the US forged austenitic stainless steel is typically used. This study focuses on fabrication QC measures required to ensure reproducible and acceptable properties for DCI containments. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗